DeSantis et al., U.S. Pub. No. 2011/0222640 A1 published Sep. 15, 2011 and incorporated herein by reference in its entirety discloses (among other subject matter) a CRDM for a nuclear reactor employing a lead screw (sometimes referred to as a ball screw herein denoting specific lead screw embodiments employing ball nuts disposed between the screw and nut threadings) engaged by a motor to provide controlled vertical translation, in which a separate latch assembly connected with the lead screw latches to the lifting rod of a control rod (or to the lifting rod of a control rod assembly comprising plural control rods connected by a yoke or spider to the lifting rod). The latch is actively closed to connect the translating assembly comprising the lifting rod and the control rod(s) so that the translating assembly translates with the lead screw under control of the CRDM motor. Upon removal of the closing force, e.g. during a SCRAM, the latch opens to release the lifting rod and SCRAM the control rod(s), while the lead screw remains engaged with the CRDM motor and does not fall. In some illustrative embodiments, the latches are actively closed by cam bars that are lifted by a hydraulic piston, solenoid, or other lifting mechanism, where each cam bar is part of a four-bar linkage that moves the cam bar horizontally in response to the lifting in order to cam the latches shut. In DeSantis et al., U.S. Pub. No. 2011/0222640 A1, the four-bar linkage is arranged such that under gravity the four-bar linkage operates to move the cam bars outward so as to release the latch.
By way of non-limiting illustrative example, FIGS. 1 and 2 correspond to drawing sheets 1 and 16, respectively, of DeSantis et al., U.S. Pub. No. 2011/0222640 A1. With reference to FIG. 1, an illustrative nuclear reactor vessel of the pressurized water reactor (PWR) type is diagrammatically depicted. An illustrated primary vessel 10 contains a reactor core 12, internal steam generator(s) 14, and internal control rods 20. The illustrative reactor vessel includes four major components, namely: 1) a lower vessel 22, 2) upper internals 24, 3) an upper vessel 26 and 4) an upper vessel head 28. A mid-flange 29 is disposed between the lower and upper vessel sections 22, 26. Other vessel configurations are also contemplated. Note that FIG. 1 is diagrammatic and does not include details such as pressure vessel penetrations for flow of secondary coolant into and out of the steam generators, electrical penetrations for electrical components, and so forth. The lower vessel 22 of the illustrative reactor vessel 10 of FIG. 1 contains the reactor core 12, which can have substantially any suitable configuration. The illustrative upper vessel 26 houses the steam generator 14 for this illustrative PWR which has an internal steam generator design (sometimes referred to as an integral PWR design). In FIG. 1, the steam generator 14 is diagrammatically shown. In a typical circulation pattern the primary coolant is heated by the reactor core 12 and rises through the central riser region 32 to exit the top of the shroud 30 whereupon the primary coolant flows back down via the downcomer region 34 and across the steam generators 14. Such primary coolant flow may be driven by natural convection, by internal or external primary coolant pumps (not illustrated), or by a combination of pump-assisted natural convection. Although an integral PWR design is illustrated, it is also contemplated for the reactor vessel to have an external steam generator (not illustrated), in which case pressure vessel penetrations allow for transfer of primary coolant to and from the external steam generator. The illustrative upper vessel head 28 is a separate component, but it is also contemplated for the vessel head to be integral with the upper vessel 26. While FIG. 1 illustrates an integral PWR, in other embodiments the PWR may not be an integral PWR, that is, in some embodiments the illustrated internal steam generators may be omitted in favor of one or more external steam generators. Still further, the illustrative PWR is an example, and in other embodiments a boiling water reactor (BWR) or other reactor design may be employed, with either internal or external steam generators.
With reference to FIG. 2, a control rod system embodiment is described, e.g. suitably part of the upper internals 24 of the nuclear reactor of FIG. 1, which provides electromagnetic gray rod functionality (i.e. continuously adjustable control rod positioning) and a hydraulic latch system providing SCRAM functionality (i.e. in an emergency, the control rods can be fully inserted in order to quickly quench the nuclear reaction, an operation known in the art as a SCRAM). The control rod system of FIG. 2 allows for failsafe SCRAM of the control rod cluster without scramming the lead screw. A motor/ball nut assembly is employed, such that a lead screw 40 is permanently engaged to a ball-nut assembly 42 which provides for axial translation of the lead screw 40 by driving a motor 44. The illustrative motor 44 is mounted on a standoff 45 that positions and bottom-supports the motor 44 in the support structure of the upper internals 24; other support arrangements are contemplated. A control rod cluster (not shown) is connected to the lead screw 40 via a lifting/connecting rod or lifting/connecting rod assembly 46 and a latch assembly 48. The lead screw 40 is substantially hollow, and the lifting/connecting rod 46 fits coaxially inside the inner diameter of the lead screw 40 and is free to translate vertically within the lead screw 40. The latch assembly 48, with spring loaded latches, is attached to (i.e. mounted on) the top of the lead screw 40. When the latches of the latch assembly 48 are engaged with the lifting rod 46 they couple the lifting/connecting rod 46 to the lead screw 40 and when the latches are disengaged they release the lifting/connecting rod 46 from the lead screw 40. In the illustrated embodiment, latch engagements and disengagements are achieved by using a four-bar linkage cam system including two cam bars 50 and at least two cam bar links 52 per cam bar 50. Additional cam bar links may be added to provide further support for the cam bar. When the cam bars 50 move upward the cam bar links 52 of the four-bar linkage also cam the cam bars 50 inward so as to cause the latches of the latch assembly 48 to rotate into engagement with the lifting/connecting rod 46. In the illustrated embodiment, a hydraulic lift assembly 56 is used to raise the cam bar assemblies 50. In an alternative embodiment (not illustrated), an electric solenoid lift system is used to raise the cam bar assemblies. When a lift force is applied to the cam system, the upward and inwardly-cammed motion of the cam bars 50 rotates the latches into engagement thereby coupling the lifting/connecting rod 46 to the lead screw 40. This causes the control rod cluster to follow lead screw motion. When the lift force is removed, the cam bars 50 swing down and are cammed outward by the cam bar links 52 of the four-bar linkage allowing the latches of the latch assembly 48 to rotate out of engagement with the lifting/connecting rod 46. This de-couples the lifting/connecting rod 46 from the lead screw 40 which causes the control rod cluster to SCRAM. During a SCRAM, the lead screw 40 remains at its current hold position. After the SCRAM event, the lead screw 40 is driven to the bottom of its stroke via the electric motor 44. When the lift force is reapplied to the cam system via the hydraulic lift assembly 56, the latches of the latch assembly 48 are re-engaged and the lifting rod 46 is re-coupled to the lead screw 40, and normal operation can resume. Other latch drive modalities are contemplated, such as a pneumatic latch drive in which pneumatic pressure replaces hydraulic pressure in the illustrated lift assembly 56. In FIG. 2, the lead screw 40 is arbitrarily depicted in a partially withdrawn position for illustration purposes. The latching assembly 48 is attached to (i.e. mounted on) the top of the lead screw 40. The ball nut 42 and motor 44 are at the bottom of the control rod drive mechanism (CDRM), the latch cam bars 50 extend for the full length of mechanism stroke, and the hydraulic lift system 56 is located at the top of the mechanism. In some embodiments, the CRDM of FIG. 2 is an integral CDRM in which the entire mechanism, including the electric motor 44 and ball nut 42, and the latching assembly 48 are located within the reactor pressure vessel 10 (see FIG. 1) at full operating temperature and pressure conditions. Further illustrative embodiments of CRDM designs employing the cam bars with four-bar linkages are described in DeSantis et al., U.S. Pub. No. 2011/0222640 A1, which is incorporated herein by reference in its entirety.